EPRI Report MRP-211, Revision 1

© 2018 Electric Power Research Institute, Inc. All rights reserved.

Kyle AmbergeEPRI

Heather MalikowskiExelon

Steve Fyfitch, Sarah DavidsaverFramatome, Inc.

NRC Public MeetingFebruary 13, 2018

EPRI Report MRP-211, Revision 1

Subsequent License Renewal

2© 2018 Electric Power Research Institute, Inc. All rights reserved.

Agenda

Time Agenda Item Presenter8:45 am Opening remarks Kyle Amberge, EPRI9:00 am Purpose and objective of meeting Kyle Amberge, EPRI9:30 am Background/history of MRP-211 Steve Fyfitch,

Framatome10:30 am Break All10:45 am MRP-211, Revision 1 irradiated

materials database and modelsSarah Davidsaver, Framatome

12:00 pm Lunch All1:30 pm Discussion of MRP-211, Revision 1 NRC Staff2:30 pm Responses to Other Questions Industry3:00 pm Break All3:15 pm Stakeholder participation All3:30 pm Summary for the day and actions NRC Staff/EPRI3:45 pm Adjourn All


Opening Remarks

Kyle Amberge, EPRI


Roadmap for MRP-227 Development

MRP-134 – Fundamental approach and frameworkMRP-156/-157 – Materials issue management tableMRP-211 – Irradiated stainless steel propertiesMRP-175 – Screening criteria for aging mechanismsMRP-135 – Constitutive models for ANSYS FE plug-inMRP-189/-191 – FMECA ranking by componentMRP-229/-230 – Engineering analysis/finite element modelMRP-231/-232 – AMP strategies assigned by componentMRP-227 – Inspection recommendations


Purpose and Objective of Meeting

Kyle Amberge, EPRI


Purpose and Objective of 2/13-14 Meeting with NRC

To interact with NRC Staff proactively for subsequent license renewal (SLR) program development– Previous similar Reactor Vessel Internals meetings for (first) LR

pertaining to MRP-175 and MRP-211 11/16/05 – MRP-175 discussion prior to publication (meeting

summary ML053270247) 5/3/06 – MRP-175 discussion after publication (meeting summary

ML061290492) 2/23/07 – MRP-211 discussion (meeting announcement

ML070390233)– Per MRP-227-A SER, the industry considers that there are no open

items for first LRTo keep NRC Staff informed of the SLR program development

and progress of the Joint EPRI MRP Reactor Internals Core Planning TeamTo obtain feedback from the NRC Staff regarding the SLR

program development and progress of the Joint EPRI MRP Reactor Internals Core Planning Team


Purpose and Objective of NRC Meeting 2/13/18

Present irradiated austenitic stainless steel material property database, including identification of updates from MRP-211, Revision 0 (2007)Present discussion of irradiated austenitic stainless steel

material constitutive models, including identification of updates and changes from MRP-211, Revision 0 (2007)Foster technical discussion with NRC

– Review and discuss other materials related questions from staff

Identify future meetings/topics and interactions with NRC


Background/History of MRP-2112005-2007

Steve Fyfitch, Framatome


Background/History of MRP-211, Revision 0

Industry met with NRC 2/23/07 to discuss the development MRP-211 (ML070390233, Proprietary information presented at that meeting)MRP-211, Revision 0 published December 2007

– Full citation:Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data—State of Knowledge (MRP-211). EPRI, Palo Alto, CA: 2007. 1015013.

– Public access in 2009Prepared by AREVA NP, with contributions from

Westinghouse, ANATECH, and EPRI– Steve Fyfitch, Peter Scott, Lionel Fournier, Robert (Bob) Gold, Joe

Rashid, Robert Dunham, Mike BurkeMRP-211, Revision 0 submitted to NRC (see

ML093020614)



MRP-211 provides data to develop the technical bases for constitutive models to be used in engineering evaluations and assessments– Data compared to MRP-135 models and adjustments made

These evaluations and assessments are used to refine the categorization and ranking of PWR RV internals component items and welds– MRP-211, Revision 0 supports (first) license renewal, i.e., 60 years– MRP-211, Revision 1 supports subsequent license renewal, i.e.,

beyond 60 years

MRP-211 provides database and general trends/available models



Approach:– Expert panel elicitation– Relevant data gathered, reviewed, and captured in database– Existing models (i.e., MRP-135) discussed and evaluated against

database– Recommendations identified for model changes (next MRP-135

revision)

MRP-211, Revision 0 report sections:– Section 1 – Introduction– Report purpose (Section 1.1)– Background (Section 1.2)– Report structure (Section 1.3)– References (Section 1.4)



MRP-211, Revision 0 report sections (cont.):– Section 2 – Data (sources, trends, gaps, and references) Tensile test data (Section 2.1) Fracture toughness data (Section 2.2) Thermal and irradiation creep/stress relaxation data (Section 2.3)Void swelling data (Section 2.4) IASCC initiation data (Section 2.5) IASCC growth data (Section 2.6)



MRP-211, Revision 0 report sections (cont.):– Section 3 – Material constitutive equations Tensile data models (Section 3.1) Fracture toughness models (Section 3.2) Irradiation-enhanced stress relaxation and creep model

(Section 3.3)Void swelling model (Section 3.4) IASCC initiation model (Section 3.5)References (Section 3.6)

– Section 4 – Summary



MRP-211, Revision 0 report sections (cont.):– Appendices (contain the actual raw data) Tensile property data (Appendix A) Fracture toughness data (Appendix B) Irradiation creep data (Appendix C)Void swelling data (Appendix D) IASCC data (Appendix E) IASCC growth data (Appendix F) Test specimen designs (Appendix G)Composition (Appendix H)References (Appendix I)



Process:– EPRI MRP reports (extensive data compilations) were published in

2001-2005 and other available literature data sources dating to 1950s were initially evaluated

– Expert panel review of the data in these reports was performed– Remaining data gaps were identified after this 2007 review (NRC Suppl. Question # 1.c)

– A summary of the data applicable to PWR internals was prepared



Data collected were compared to constitutive models in MRP-135, Revision 0Updates and changes made as necessary to models to fit

dataMaterial properties were characterized as a function of

neutron fluence and temperature to the extent possible–Fracture toughness, irradiation-enhanced stress

relaxation/creep, and IASCC crack initiation data were characterized using a lower bound approach

–Limitations were identified–Precision of models limited by breadth of database

MRP-135 was updated to Revision 1 (2010) incorporating the MRP-211, Revision 0 model recommendations


MRP-211, Revision 1 Irradiated Materials Database and Models

2016-2017Sarah Davidsaver, Framatome


MRP-211, Revision 1 Irradiated Materials Database and ModelsNRC Staff requested MRP-211, Revision 1 (and MRP-175,

Revision 1) on November 29, 2017 [ML17307A156]– To support SLR implementation of MRP-227, Revision 2

EPRI transmitted proprietary and non-proprietary versions of both documents on December 18, 2017 [ML17361A187]– MRP-227 roadmap, developed in 2010, was also included for

referenceThis presentation summarizes the current state-of-the-

technology of neutron irradiation-induced property changes in austenitic stainless steels and recommended degradation models provided in MRP-211, Revision 1– Comparisons are made using MRP-211 Rev.0 figures and changes

More detailed discussions of age-related degradation mechanisms (ARDMs) are in MRP-175



Approach:– Expert Panel elicitation– Relevant new data sources since 2007 publication of

MRP-211, Revision 0 identified and gathered Environmental-Degradation Conferences Fontevraud Conferences Other literature sources (e.g., Journal of Nuclear Materials) EPRI MDM and IMT documents NRC NUREG/CR and PMDM documents EPRI Materials Handbook Recent Framatome/Westinghouse evaluation reports PWR Owners Group reports ICG-EAC meetings EPRI BWRVIP, MRP and PSCR reports and meetings



Approach (cont.):– Critically review and analyze the most recently available irradiated

material data– Critically assess data fit to available models Identify alternative formulations recommended and reach

consensus, as appropriate– Identification of remaining gaps in database for potential future

actions{NRC Suppl. Question # 1.c}



MRP-211, Revision 1 published October 2017– Full citation:

Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models, and Basis Data—State of Knowledge (MRP-211, Revision 1). EPRI, Palo Alto, CA: 2017. 3002010270.

– EPRI Proprietary document

Prepared by AREVA NP, with contributions from Westinghouse, SIA (formerly ANATECH), Vattenfall, EDF, Dominion, Southern Nuclear, Exelon, and EPRI– Steve Fyfitch, Sarah Davidsaver, David Burak, Daniel Brimbal, Josh

McKinley, Randy Lott, Michael Burke, Michael Ickes, Greg Troyer, Ryan Hosler, Joe Rashid, Nathan Capps, Pal Efsing, Faiza Sefta, Jean-Paul Massoud, Glenn Gardner, Tim Wells, Heather Malikowski, Kyle Amberge, Jean Smith, Cem Topbasi, Peter Chou



Selection of age-related degradation mechanisms (ARDMs)

– Based on EPRI’s Materials Degradation Matrix (MDM)

– Non-irradiated degradation modes addressed in MRP-175: SCC (IG/TG) Wear Fatigue (HC) Reduction in Fracture

Properties (Th)

– Others addressed in MRP-211

– General corrosion mechanisms are not pertinent to austenitic stainless steels due to PWR water chemistry controls

EPRI Materials Degradation Matrix, Revision 3. EPRI, Palo Alto, CA: 2013. 3002000628.



Comparison between MRP-175 ARDMs and MRP-211 database/models

– MRP-211 contains data used for MRP-135 constitutive irradiation effects model development

– ARDMs not requiring irradiated property constitutive models, such as SCC, wear, HC fatigue, and reduction in fracture properties (Th) are not included in MRP-211

ARDM MRP-175 MRP-211

SCC Appendix A Not irradiated property

IASCC Appendix B Section 2.5 and Appendix E (Initiation) Section 2.6 and Appendix F (Growth)

Wear Appendix C Not irradiated property

Fatigue Appendix D Section 2.8 and Appendix H

Thermal Embrittlement

Appendix E Section 2.7 and Appendix G

Irradiation Embrittlement

Appendix F Section 2.1 and Appendix A (Mechanical Properties)Section 2.2 and Appendix B (Fracture Toughness)

Void Swelling Appendix G Section 2.4 and Appendix D

Stress Relaxation and

Irradiation Creep

Appendix H Section 2.3 and Appendix C



MRP-211 summarizes the available data that describes the current state-of-knowledge of neutron irradiation-induced property changes in austenitic stainless steels, principally:– Solution-annealed Type 304 and 304L– Cold-worked Type 316 and 316L– Grades CF3/CF3M and CF8/CF8M cast austenitic stainless steels– Austenitic stainless steel weld metals (e.g., Type 308)

Age-related degradation mechanisms include: Irradiation embrittlement (IE) Thermal and Irradiation-enhanced stress relaxation/creep (ISR/IC)Void swelling (VS) Irradiation-assisted stress corrosion cracking (IASCC) Fatigue including EAF



MRP-211, Revision 1 supports:– Evaluations and assessments required to refine/update

categorization and ranking of PWR internals components and items– SLR I&E guideline (i.e., MRP-227, Revision 2) development

Update to MRP-211 used same process that generated Rev.0Updated MRP-135 constitutive models will be based on

recommendations in MRP-211, Revision 1– Revision scheduled for mid-2018



Fundamental differences between MRP-211, Revision 1 and MRP-211, Revision 0:– Section 1 – IntroductionUpdated each sub-section, as appropriate

– Section 2 – Data (sources, trends, gaps, model, and references) Tensile test data (Section 2.1) Fracture toughness data (Section 2.2) Thermal and irradiation creep/stress relaxation data (Section 2.3)Void swelling data (Section 2.4) IASCC initiation data (Section 2.5) IASCC growth data (Section 2.6)Combined TE and IE data (Section 2.7) Fatigue (including EAF) life data (Section 2.8)


MRP-211, Revision 1 Irradiated Materials Database and Models Fundamental differences between MRP-211, Revision 0 and MRP-211, Revision 1 (cont.):

– Section 3.0 – Summary Previous section 3.0 was completely revised and model recommendations were

incorporated into current Section 2.0– MRP-211, Revision 0 Section 4.0 – Summary No longer required in MRP-211, Revision 1

– Appendices - tables updated with new data Tensile property data (Appendix A) Fracture toughness data (Appendix B) Irradiation creep data (Appendix C) Void swelling data (Appendix D) IASCC data (Appendix E) IASCC growth data (Appendix F) TE and IE data (Appendix G) Fatigue (including EAF) data (Appendix H) Test specimen designs (Appendix I) Composition (Appendix J) References (Appendix K)


Break



Irradiated Materials Database






MRP-211, Revision 1 Irradiated Materials Database and Models (Tensile)

Tensile properties– Irradiation produces defects and

precipitates that form obstacles to dislocation movement, which causes: an increase in yield strength (YS) and

ultimate tensile strength (UTS) a decrease in uniform elongation (%UE)

and total elongation (%TE)– Tensile properties of irradiated stainless

steels have been widely investigated Fast reactor irradiations Thermal reactor irradiationsData trends are consistent

– Majority of available data from fast reactors; however, significant LWR data are now available for comparison and evaluation



Tensile properties–Data compiled in Section 2.1 and Appendix A–Figures 2-2 through 2-30 show dataRoom and high temperature, all properties

–28 additional references addedEPRI MRP and BWRVIP reports, conference proceedings, journals, PhD thesis, Halden reactor project reports, NUREG/CR reportData has been extended from 124 dpa to 231 dpa

Final conclusion is unchanged from MRP-211, Rev. 0– Data demonstrate that tensile properties saturate

by a neutron exposure of ~20 dpa (~1.33E22 n/cm2, E > 1 MeV) (answers NRC suppl. question #1.b)



The Effect of Neutron Fluence on Room Temperature Yield Strength for Solution-Annealed and Cold-Worked Type 316, Type 347, and Type 348 Stainless Steels (MRP-211 Rev. 0 Figure 2-2)

MRP-211, Revision 1 contains:– Significantly more

data in the 0-20 dpaand 140-231 dparanges

– Additional data to fill in gaps in 30-50 dparange

– Data obtained from: Fast reactors and

thermal reactors Low and high test

temperatures– These data will be

used in the current MRP-135 models

Example Rev.0 Figure


MRP-211, Revision 1 Irradiated Materials Database and Models (Fracture Toughness)

Fracture toughness– Irradiation produces defects and

precipitates, which causes a reduction of fracture toughness in austenitic stainless steels

– The fracture toughness of stainless steel saturates when exposed to high-energy neutrons

– The reduction of fracture toughness with increasing neutron dose in both BWRs and PWRs is consistent with that observed in fast reactors

– Majority of available data from fast reactors; however, significant LWR data are now available for comparison and evaluation


MRP-211, Revision 1 Irradiated Materials Database and Models (Fracture Toughness)

Fracture toughness properties– Data compiled in Section 2.2 and Appendix B– Figures 2-31 through 2-34 show data– 16 additional references added Conference proceedings, journals,

NUREG/CR, EPRI BWRVIP and MRP letter and report, NRC final safety evaluation New data mainly in the 0 to 15 dpa range

Fracture toughness data with VS up to ~8% is similar to other data at lower doses without significant swelling (NRC suppl. question # 5.b)

The conclusion is unchanged from MRP-211, Rev. 0– All data remain bounded by a

saturated value of 38 MPa√m (34.6 ksi√in) for fluence greater than about 10 dpa (~6.67E21 n/cm2, E>1 MeV)

Example Rev.0 Figures


MRP-211, Revision 1 Irradiated Materials Database and Models (Thermal and Irradiation Creep/Stress Relaxation)Thermal and irradiation stress

relaxation/creep– Thermal stress relaxation (creep) occur

as a result of mobile vacancies, dislocation climb, and a material’s defect structureConventionally, these occur above

~0.5Tm– However, under irradiation conditions

such processes can occur at lower temperatures

– ISR/IC depends mainly on the neutron fluence and the applied stress

– Manifests itself in relaxation of preloaded components (potentially leading to excessive wear or fatigue)Evident when BFBs are removed

(based on removal torque)


MRP-211, Revision 1 Irradiated Materials Database and Models (Thermal and Irradiation Creep/Stress Relaxation) Thermal and irradiation creep/stress

relaxation (ISR/IC) properties– Data compiled in Section 2.3 and Appendix C– Figures 2-39 and 2-40 show data– 14 additional references added

Conference proceedings, journals, EPRI MRP report, EdF report

– Creep compliance values [linear correlation of (effective creep strain) with (effective stress * irradiation dose)] slightly changed by new data Dashed line versus solid line change is

insignificant

The conclusion is unchanged from MRP-211, Rev. 0

– Correlations still indicate that a greater creep rate occurs for Type 304 SA material than for Type 316 CW material

Example Rev.0 Figures


MRP-211, Revision 1 Irradiated Materials Database and Models (Void Swelling)

Void swelling– Irradiation causes displacement of atoms from

their lattice sites leading to formation of cavities (or, voids), which causes: Volume and dimensional changes Potential distortions of structural components Potential for a reduced tearing modulus at

PWR operating temperatures that may fall to zero at room temperature

– VS is strongly sensitive to dose, dose rate, irradiation temperature, and material composition

– Fast reactor data are insufficient to estimate VS trends in LWRs because different temperatures, damage rates, and helium production rates

– PWR irradiation spectrum data show very little void swelling although data are limited



Void swelling properties– Data compiled in Section 2.4 and Appendix D– Figure 2-41 shows data– 32 additional references added Conference proceedings, journals, EPRI MRP

reports, ORNL report, PhD thesis – A cluster-dynamics-based VS model was

developed through another EPRI-sponsored project and is recommended for calculation of VS in PWR environments (MRP-391)

This model is added to MRP-211, Rev. 1– The VS model indicates that steady-state swelling

rates of less than 0.1 %/dpa are reasonable for the fluence levels and temperatures expected in PWR internals during SLR

– Data of extracted PWR internals components and/or materials are in agreement with the cluster-dynamic model



Swelling Predictions vs. dpa for Solution Annealed 300 Series Stainless Steels at 320°C (608°F) at a Helium-to-dpa ratio of 10 appm He/dpa, Various Displacement Rates Relevant to a PWR Environment (Davidsaver et al., 2017 Environmental Degradation Conference)

(NRC Suppl. Question 5.a and 5.b)

Typical LWR conditions


MRP-211, Revision 1 Irradiated Materials Database and Models (IASCC Initiation)

IASCC initiation properties– IASCC results from a combined

effect of irradiation damage to the material, stress state and water environmentNo single mechanism that controls IASCC initiation has been identified

– IASCC initiation data show both initiation and no-initiation domains



IASCC initiation properties– Data compiled in Section 2.5 and Appendix E– Figures 2-45 through 2-47 show data– 19 additional references addedConference proceedings, journals, Halden reports,

NUREG/CR– Laboratory test data indicate that IASCC initiation

susceptibility appears to continue to increase with irradiation damage, even though the tensile properties appear to saturate by 20 dpa (~1.33E22 n/cm2 , E > 1.0 MeV),

Change made from MRP-211, Rev. 0– IASCC crack initiation may not occur in materials irradiated to about 80

dpa (~5.33E22 n/cm2 , E > 1.0 MeV) when component loaded to below approximately 35% of irradiated yield strength Lower bound trending model adjusted



IASCC Flaw Initiation Stress Versus Dose – Constant Load Tests (MRP-211 Rev. 0 Figure 2-38)

Source: Fyfitch et al., 2017 Environmental Degradation Conference

IASCC initiation and no-initiation domains

NRC Suppl. Question # 3Initiation = above trend line

No-Initiation = below trend line


MRP-211, Revision 1 Irradiated Materials Database and Models (IASCC Growth)

IASCC growth properties– Subcritical crack growth by IASCC

occurs as a result of three factors: Stress intensity Neutron fluence level Temperature

– IASCC Expert Panel to gather, review, and evaluate crack growth data for both PWRs and BWRs

– Data then used to create recommended CGR disposition curves

– CGR curves were submitted to the ASME Code in a Code Case for use by the industry

– Core barrel welds expected to be well below maximum fluence of the CGR dataset {NRC Suppl. Question #4}


MRP-211, Revision 1 Irradiated Materials Database and Models (IASCC Growth)

IASCC growth properties– Data compiled in Section 2.6 and Appendix F– Figures 2-48 through 2-49 show example disposition lines– All references removed, only one reference used Models of Irradiation-Assisted Stress Corrosion Cracking of

Austenitic Stainless Steels in Light Water Reactor Environments: Volume 1: Disposition Curves Development and Volume 2: Disposition Curves Application. EPRI, Palo Alto, CA: 2014. 3002003103.

– Disposition curves based on stress intensity factor, KApplicable to flaw evaluations in PWR and BWR plants

This is an addition to MRP-211, Rev. 1


MRP-211, Revision 1 Irradiated Materials Database and Models (IASCC Growth) Example IASCC crack growth rates (PVP2015-45323)


MRP-211, Revision 1 Irradiated Materials Database and Models (Combined TE and IE of CASS)

TE and IE properties– The combined effect of TE and IE is a

time and dose dependent process whereby a material undergoes microstructural changes leading to decreased ductility and degradation of toughness and impact properties

– CASS is a two-phase material consisting primarily of an austenite matrix with the remainder being ferrite

– Both the austenite and the ferrite are affected by irradiationOnly the ferrite is affected by

temperature– Limited data are available on the

combined effects of these two embrittlement mechanisms


MRP-211, Revision 1 Irradiated Materials Database and Models (Combined TE and IE of CASS)

TE and IE properties– Data compiled in Section 2.7 and Appendix G (New in Revision 1)– Figure 2-50 shows data– Data from NUREG/CR reports and one paper from conference proceedings Very limited data

– Differences observed in material types, testing environments, and temperature CF3, CF3M, CF8, and CF8M Air versus water environment Room temperature versus service temperature

– CASS are shown to be susceptible to loss of toughness by combined thermal embrittlement and IE, which is shown to depend on the extent of the ferrite phase Assessments of CASS have shown that PWR reactor internals components

are not significantly impacted by TE/IE (Ref. ML16250A001)No corresponding section in MRP-211, Rev. 0

– A lower bounding curve has been developed for MRP-211, Rev.1


BREAK


MRP-211, Revision 1 Irradiated Materials Database and Models (Fatigue Life)

Fatigue properties– Fatigue is a process involving the

evolution of persistent slip bands at the surface of a material and subsequent crack formation and propagation during exposure to cyclic stresses

– Low-cycle fatigue (LCF) is associated with plastic strains

– High-cycle fatigue (HCF) occurs at stresses below the elastic limit

– For both cases, the environment can impact the final fatigue results and this is known as environmentally-assisted fatigue (EAF)


MRP-211, Revision 1 Irradiated Materials Database and Models (Fatigue Life)

Fatigue life properties– Data compiled in Section 2.8 and Appendix H (New in Rev. 1)– Figures 2-51 through 2-53 show data– Data from NUREG/CR-6909 report, conference proceeding,

and EPRI MRP report

No corresponding section in MRP-211 Rev. 0– The expert panel recommended applying existing methods for

evaluating fatigue life on irradiated materials with a suggested environmental correction in accordance with NUREG/CR-6909, Revision 1As more test data are gathered, this approach may be

updated



Irradiated Materials Models



MRP-211, Revision 1 major changes:

– Void swelling – Updated to incorporate cluster dynamics-based model

– Irradiation creep – Change in creep compliance terms and removal of incubation period

– IASCC Initiation – Updated to be consistent with the conservative fit to the database

– IASCC Growth – Use of 2014 industry developed disposition curves– TE & IE (new) – Bounding trend line developed for available CASS

data– Fatigue life (new) – Based on NUREG/CR-6909 Revision 1

environmental correction factors


Summary and Conclusions

MRP-211, Revision 0 was published in 2007– Summarized current knowledge of irradiated stainless steel properties– Provided recommended models for changes to MRP-135 constitutive

models– Used to refine screening and categorization results through engineering

evaluations and assessmentsAdditional testing and operating experience data from the last

~10 years since publication have been gatheredMRP-211, Revision 1 was published in 2017

– Updated the database to include recent data, address gaps as applicable, update models

– Include new sections for TE + IE data and fatigue (including EAF) data– Recommended model changes including updates and new models

Revision 1 models will be used in developing MRP-227, Revision 2 for SLR and through a revision to MRP-135, Revision 1


Summary and Conclusions

Expert panel validated the 2007 materials property assessments in MRP-211 Revision 0– Additional testing and operating experience data from the last ~10

years confirm many of the same conclusions in MRP-211, Revision 0 – Additional testing and operating experience data from the last ~10

years provide improvements to MRP-135 models due to a more extensive database of neutron-induced ARDMs

– Data for two additional ARDMs (combined TE + IE) and fatigue (including EAF) were added to the database

Industry provided MRP-211 Revision 1 to NRC for info to foster continued technical exchange with staff


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